High Integrity Container (HIC) made of polymer concrete was developed for the efficient treatment and safe disposal of radioactive spent resin and concentrate waste in consideration of the disposal requirements of domestic disposal sites. Permission for application of Polymer Concrete High Integrity Container (PC-HIC) to the domestic nuclear power plants (NPPs) has been completed or is under examination by the regulatory agency. In the case of 860 L PC-HIC for very-low-level-waste (VLLW) or low low-level-waste (LLW), the application of four representative NPPs has been approved, and the license for extended application to the rest NPPs is also almost completed. A licensing review is also underway to apply 510 L PC-HIC for intermediate and low-level-waste (ILLW) to representative nuclear power plants. In order to handle and efficiently store and manage PC-HICs and high-dose PCHIC packages, a gripper device that can be remotely operated and has excellent safety is essential, and the introduction of NPPs is urgent. The conventional gripper device developed by the PC-HIC manufacturer for lifting test to evaluate the structural integrity of PC-HIC requires a rather wide storage interval due to its design features, and does not have a passive safety design to handle heavy materials safely. In addition, work convenience needs to be reinforced for safety management of high radiation work. Therefore, we developed a conceptual design for a gripper device with a new concept to minimize the work space by reflecting on-site opinions on the handling and storage management conditions of radioactive waste in NPPs, and to enhances work safety with the passive safety design by the weight of the package and the function of checking the normal seating of the device and the normal operation of the grip by the detector/indicator, and to greatly improves the work efficiency and convenience with the wireless power supply function by rechargeable battery and the remote control function by camera and wireless monitoring & control system. Through design review by experts in mechanical system, power supply and instrumentation & control fields and further investigations on the usage conditions of PC-HICs, it is planned to facilitate preparations for the application of PC-HIC to domestic NPPs by securing the technical basis for a gripper device that can be used safely and efficiently and seeking ways to introduce it in a timely manner.

Encapsulation using cement as a solidification method for disposal of radioactive waste is most commonly used in most advanced countries in the nuclear technology to date due to its advantages such as low material cost and accumulated technology. However, in case of cement solidification, since moisture or hydroxyl group in cement is decomposed by radioactivity, it may happen that gas is generated, structural stability is weakened, and leachability is increased due to low chemical durability. Therefore, the various new solidification methods are being developed to replace it. As one of these alternative technologies, for dispersible metal compounds generated through the incineration replacement process, the study on engineering element technology for powder metallurgy is under way, which overcomes the interference problem between mechanical elements and media that may occur during the process such as the homogeneous mixing process of the target powder substance and additives used in the powder metallurgy concept-based sintering process for the solidification of the final glass composite material (GCM), the process of creating a compressed molded body using a specific mold, the process of final sintering treatment. The solidification process of dispersible radioactive waste can be largely divided into pre-treatment stage, molding stage, and sintering stage, and the characteristics of the final radioactive waste solidification material can vary depending on the solidification treatment characteristics of each stage. In relation with these characteristics, the matters to be considered when designing device for each stage to solidify dispersible radioactive waste (property of super-mixing device for homogenized powder formation, structural geometry and pressure condition of molding device for production of compressed molded body, temperature and operation characteristics of sintering device for final glass composite material (GCM), etc.) are drawn out. It is expected that the solidification device design reflecting these considerations will meet all disposal conditions of radioactive waste material, such as compressive strength and leaching characteristics of solidified radioactive waste material, and create a uniformized solidification of radioactive waste material.

Glass wool, the primary material of insulation, is composed of glass fibers and is used to insulate the temperature of steam generators and pipes in nuclear power plants. Glass fiber is widely adopted as a substitute for asbestos classified as a carcinogen. The insulations used in nuclear power plants are classified as radioactive waste and most of the insulation is Very Low-Level Waste (VLLW). It is packaged in a 200 L drum the same as a Dry Active Waste (DAW). In the case of the insulations, it is packaged in a vinyl bag and then charged into the drum for securing additional safety because of the fine particle size of the fiberglass. A safety assessment of the disposal facility should be considered to dispose of radioactive waste. As a result of analyzing overseas Waste Acceptance Criteria (WAC), there is no case that has a separate limitation for glass fiber. Also, in order to confirm that glass fibers can be treated in the same manner as DAW, research related to the diffusion of glass fibers into the environment was conducted in this paper. It was confirmed that the glass fiber was precipitated due to the low flow velocity of groundwater in the Gyeongju radioactive waste repository and did not spread to the surrounding environment due to the effect of the engineering barrier. Therefore, the glass fiber has no special issue and can be treated in the same way as a DAW. In addition, it can be disposed of in the disposal facility by securing sufficient radiological safety as VLLW.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

Following a radioactive waste criterion and clearance level radioactive waste Act Article 2. “The radioactive wastes confirmed by the Commission as having concentration by nuclide not exceeding the value determined by the Commission through incineration, reclamation, recycling, etc”. The combustible clearance level radioactive wastes like lumbers are incinerated and non-combustible wastes like concreted are buried. The metals clearance level radioactive wastes are recycled after being re-molded. However, the clearance level radioactive waste with keeping its original forms is not common. Due to the nature of KAERI, the equipment are brought into the radiation-controlled zone for experiments. Those equipment are conservatively considered contaminated and categorized with radioactive waste following nuclear safety acts. In this case, the spectroscopy device which is clearance level radioactive waste is self-disposed for use in non-controlled areas. The 4 devices are composed of 3 gamma-ray spectroscopy and 1 alpha, beta counting system. Those devices were used for clearance level radioactive waste’s radioisotope analysis in Radioactive Waste Form Test Facility which is used in a separated room for analysis. This room will be released in nonradiation controlled area, therefore those devices will be moved to non-controlled area and keep using. Last April self-disposal was reported to the regulatory body and got acceptance last May. Those devices were moved to non-controlled area last July. This case will be good example for reuse equipment which stop using in radiation controlled area but can keep used.

The Nuclear Cycle Experiment Research Center is one of the facility of the Korea Atomic Energy Research Institute (KAERI). This facility is a laboratory-scale version of pyro-processing technology. Mixture depleted Uranium (DU) and depleted Uranium (DU) feed material are used in this facility for pyro-research. During summer, air conditioners that maintain temperature and humidity are always in operation to protect analysis equipments. 15 air conditioners are installed in this facility. The condensate which is generated in 15 air conditioners is collected in one place to analyze. Sampling was performed to check the level of contamination, U, pH and gamma radiation test were performed. This paper shows the degree of contamination of air conditioner condensate which is generated in the radiation management area.

Cellulose-based wastes can be degraded into short-chain organic acids at the cementitious radioactive waste repository. Isosaccharinic acid (ISA), one of the main degradation products, can form the chelate complex with metals and radionuclides, and these complexes have a potential that can accelerate to move the radionuclides to far-field from the repository. This study characterized the amount of generated ISA from typical cellulosic materials in the repository. Two different degradation experiments were conducted under alkaline conditions (saturated with Ca(OH)2 at pH 12.4): i) cellulosic material mixture under an opened condition (partially aerobic), and ii) cellulosic material under an anaerobic condition in a nitrogen-purged glove box. In the first case, three different types of cellulosic materials–paper, cotton, and wood– were mixed at the same ratio, and the experiments were carried out at three different temperatures (20°C, 40°C, and 60°C). It revealed that both the cellulose degradation rate and generated ISA concentration were high at high reaction temperatures, and various soluble degradation products such as formic acid and lactic acid were generated. The cellulose degradation in this work seems to still stay at a peeling-off process. In the second study, each type of cellulosic material was applied in its own batch experiments, and the amount of generated ISA was in the order of paper > wood > cotton. The above two experiments are supposed to be a long-term study until the generated ISA reaches an equilibrium state.

Colloid migration is an important topic in post-closure safety assessment of radioactive waste repository as radionuclide can be adsorbed onto colloidal particles and migrated along with the colloids. This would reduce retardation of radionuclide migration, thus increasing the released concentration into biosphere. Recently, glass fiber waste has been found to contain small sized crushed glass fiber particles (GFPs), and concerns regarding the colloidal impact of GFP is being discussed. In this study, relevance of assessing GFPs facilitated radionuclide transport in the disposal environment of 1st phase disposal facility. Colloidal impact assessment can be divided into two sections, colloid mobility, and colloid sorption assessments. Considering GFP being denser than water, fluid velocity of 1st phase disposal facility is too slow to initiate movement of such dense particles. GFPs would remain settled, and no colloidal impact is expected. In this study, sorption assessment mainly focused to analyze the possible impact if migration of GFP does occur. The GFP is mainly composed of SiO2 and few other metal oxides. Due to high composition of SiO2 in the GFPs, negative surface charge is induced onto the surface of the GFPs in alkaline environment. This negatively charged surface can attract free positive ions (ex. Ni, Co, Fe, etc.) in the repository, and these ions would be adsorbed onto the surface of the GFPs via coulomb force. Thus, if GFPs migrate, colloid facilitated radionuclide transport can be expected. However, before being released into the biosphere, particles must pass through the engineered and natural barriers, where ion-colloid-rock interactions could result in transfer of radionuclide from one media to another. At Naka Research Center, Japan, ion-colloid-rock interactions are experimented with bentonite colloid, and the result showed that despite colloid’s sorption ability was 10 times higher than the barrier material, the overall released radionuclide concentration has negligible change. To reflect such phenomenon, coulomb attractive force of GFPs and concrete is calculated and compared, which the result showed that glass fiber was 10 times weaker than concrete. Considering the Japan’s experimental result, glass fiber facilitated transport would not enhance the radionuclide release into the biosphere. Nonetheless, assuming GFPs being mobile in 1st phase disposal facility, GFPs’ sorption ability is found to be negligible compared to the concrete of the repository, thus radionuclide transport is not expected to be enhanced. In future, this study could be used as basis for further colloidal impact analysis for the safety assessment of the repository.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

Currently, Hanul NPP packages glass fiber classified as particulate waste in plastic packaging bags and stores them in 200 L drums. KORAD’s Waste Acceptance Criteria (WAC) presents that very low-level soil can be immobilized by loading it in a soft bag and then packaging it in a 200 L or 320 L steel drum. As currently accepted method of packaging with soft bag applies to only very low-level soils among the wastes with a risk of dispersion, it is necessary to develop a non-dispersible treatment suitable for the characteristics of other particulate waste in the future. Therefore, in order for Hanul packaging pack to be approved as an alternative method for immobilization of dispersible substances, it is necessary to verify the suitability of the packaging bag. In this paper, whether the glass fiber packaging bag used in Hanul NPP satisfies the characteristic of the soft bag presented in the WAC and the possibility of being considered as a non-dispersible measure for particulate are examined. The soft bag must meet the following requirements: material and structure, shape, drop test, and immersion test. The results of the review are as follows. First, since the glass fiber is already packaged in the drum, only the role of the inner layer, made of polyethylene, having a watertight function may be required. Second, when packaging a drum, the packaging bag is compressed into a shaped frame having an inner size of a 200 L drum, so it is packaged with little empty space in the drum. Third, as a result of a drop test of a packaging pack containing 20 kg of contents from a height of 1.2 m, it was confirmed that there was no leakage of contents. Fourth, the packaging bag was immersed in a 1-m depth water tank for 30-minutes, and the performance corresponding to the IPX7 was satisfied. As a result of reviewing the soft bag characteristic of Hanul glass fiber packaging bag, it is considered that the bag can be used as one of the non-dispersible measures because it meets almost the characteristics required by the WAC. In addition, the acceptance criteria of overseas disposal sites present various secure packaging methods in place of immobilization as a non-dispersible measure for waste containing particulate matter. It is necessary to reflect these overseas cases in the establishment of non-dispersible measures for domestic waste acceptance in the future.

The number of nuclear power plants that are permanently shut down or decommissioned is increasing worldwide, and accordingly, research is being conducted on an appropriate method for disposing of radioactive waste generated during the decommissioning of nuclear power plants. In the case of waste liquid generated during the decommissioning of nuclear power plants, it is important not only to efficiently reduce waste but also to secure the suitability of disposal. One of the solidification treatment methods for radioactive waste is cement solidification, but since cement solidification has poor solidification properties and generates a large amount of waste, improvement activities have been pursued. This study aims to develop high-performance cement-based materials and solidification treatment technology for solidification of liquid radioactive waste generated during nuclear decommissioning in order to improve the problems of cement solidification treatment method. For the development of polymer cement, epoxy resin and polyamine/amide mixed type and general Portland cement were mixed in various ratios. The most appropriate mixing ratio was 4.5:2, which showed the highest compressive strength. A simulated waste liquid was prepared by referring to the preliminary decommissioning plan of Shin-Kori Units 5 and 6, and it was dried and made into granules. Polymer cement was injected into a drum filled with granules by vacuum pressure to prepare a waste form matrix. In the solidification process, granules made by drying the waste liquid were used, and the solidification agent was filled in between the granules, so the total volume of solid radwaste was reduced compared to the conventional cement solidification treatment method. As a result, the amount of waste decreased to about 1/3, and the volume reduction rate increased by about 2.2 times. The compressive strength of 3,243 psi was confirmed in the disposability performance test for the manufactured solid samples. The compressive strength after the thermal cycling test, irradiation test, microorganism test, and immersion test was 2,257 psi, 2,306 psi, 4,530 psi, and 2,263 psi, respectively, exceeding the acceptance criteria of 500 psi. The leaching index was 7~13, and no free standing water was generated.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

The integrity of the disposal repository structure must be guaranteed for few hundreds to few hundred thousand years until toxicity of radioactive waste is surely degraded. Acoustic emission (AE) method is widely utilized to evaluate the integrity of the structure because it can detect crack wave signals of the structures. It is well known that the cracking AE energy is proportional to the volume of the structure (Fractal theory). However, it is hard to destroy whole structures for obtaining AE energy. Therefore, the scaled specimens are prepared to obtain the relationship between volume of the structure and AE energy. The specimens are prepared with same of Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC) silo concrete recipe. Their diameters are from 50 mm to 150 mm in each 10 mm and their heights are twice of the diameter. One set of 50 mm to 150 mm specimens (11 specimens in one set) are made in single mixers to maintain uniformity. Surface of the specimens are flatten with cement milk to prevent from applying load with eccentricity. The uniaxial compression test is performed by controlling displacement as 0.1 mm/min. The fractal constant is obtained using least square function from volume-cumulative AE energy relationship.

The high-level nuclear waste (HLW) repository is a 500-1,000 m deep geological disposal system with a very long life expectancy for disposing of high-level waste, which is known to have a half-life of several thousand years. This repository is subject to harsh environmental conditions, such as high temperature and radiation from high-level waste, that can cause deterioration and crack. When radiation escapes through cracks, it can injure persons on the ground. Therefore, it is essential to install a sensor that can detect problems such as cracks. But, since the high-level nuclear waste (HLW) repository is sealed with bentonite and backfill, the sensor cannot be removed or replaced once it has been installed. Therefore, it is necessary to develop a highly durable monitoring sensor that can withstand harsh environmental conditions. Before attempting to improve durability, it is first required to assess durability quantitatively. And an accelerated life test is a widely used method for assessing durability. However, it is important to obtain the same failure mode when conducting a reliability test, such as an accelerated life test. If the accelerated life test is conducted using different failure modes, the dependability of the results is inevitably diminished. Therefore, in this study, a representative failure mode for the piezoelectric sensor used in the accelerated life test was derived through experiments and literature research.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

Backfill is one of the main components of engineered barrier in a high-level waste repository. The material selection of the backfill determines the barrier performance of the backfill. Overseas, its related research has been carried out mainly in Sweden, Finland, Canada, and Japan. However, Korea has recently started backfill research, and it is urgent to select a potential material for establishing the concept of backfill material and conducting backfill research. This study reviews NEA report, potential materials for overseas backfill research, advantages and disadvantages of single and mixed backfill materials, cases of license applications in Finland and Sweden for the selection of potential materials for backfill in Korea’s high-level waste repository. The review results indicated that it is reasonable to carry out backfill research according to the following plan: Both single and mixed materials are considered as potential materials for backfill research; experiments and performance studies are conducted with these materials; and, based on the results, a potential material or candidate material for the backfill suitable for the HLW repository in Korea is determined. For this plan, the single material is tentatively selected, as in Sweden, as bentonite with a montmorillonite content of about 40-50%. Then, if the selection criteria for montmorillonite content are determined through experiments and performance studies, we determine the final potential backfill material. As for the mixed backfill material, the bentonite/crushed rock mixture seems to be more advantageous than the bentonite/sand mixture considering the disposing problem of crushed rock generated from tunnel excavation and economic feasibility through its recycling. It is thought that the bentonite used in the bentonite/crushed rock mixture should have a higher montmorillonite content than bentonite used as a single backfill material since the crushed rock acts as an inert material in the mixture. The results of this study can be used as basic data for selecting the backfill material to be applied to the high-level waste repository in Korea, and can be used as a guideline for selecting the potential material required for backfill experiments and performance studies to be carried out in the future.

The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Compacted bentonite buffer materials are a key component of the engineered barrier system for high-level radioactive waste disposal. The bentonite buffer is saturated via groundwater flow through the excavation damaged zone in the adjacent rock mass. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities. Groundwater flow can cause bentonite erosion and transportation of bentonite colloids. This bentonite mass loss can negatively impact the long-term integrity of the engineered barrier system. Hence, it is necessary to understand the effects of erosion on the properties of the bentonite buffer. In this study, a series of artificial fracture erosion experiments are conducted to investigate the erosion characteristics of compacted Ca-bentonite buffer materials for different initial dry density conditions. Compacted bentonite blocks and bentonite pellets were manufactured using the cold isostatic pressing technique and granulation compactor respectively. The specimens were placed in a custommade transparent artificial fracture cell and the bentonite intrusion characteristics were monitored for two months under free swelling conditions with no groundwater flow. The radial expansion of the bentonite specimens within the artificial fracture was measured using a digital camera. In addition, the swelling pressure, displacement, and saturation were determined using a load cell-piston system, LVDT, and electrical resistivity electrodes respectively. A hydro-mechanical-chemical coupled dynamic bentonite diffusion model was applied to model the bentonite erosion characteristics using COMSOL Multiphysics.